23808 Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 

Agency: Employment and Training 

Administration. 

Type of Review: Extension without 

change of a currently approved 

collection. 

Title of Collection: Work Application/ 

Job Order Recordkeeping. 

OMB Control Number: 1205–0001. 

Agency Form Number: N/A. 

Affected Public State, Local, or Tribal 

Governments. 

Total Estimated Number of 

Respondents: 52. 

Total Estimated Annual Burden 

Hours: 416. 

Total Estimated Annual Costs Burden 

(Operation and Maintenance): $0. 

Description: Work applications 

(commonly referred to as the 

registrations) are used in One-Stop 

Career Centers for individuals seeking 

assistance in finding employment or 

employability development services. 

They are used to collect information 

such as: applicants’ identification, 

qualifications, work experience, and 

desired pay. They also include services 

provided to applicants, such as job 

development, referral to supportive 

service. 

Job orders are used in One-Stop 

Career Centers to obtain information on 

employer job vacancies. Information in 

the job orders include employer 

identification, job requirements, pay 

information as well as identification of 

persons referred, hired, or refused. The 

information is collected at the 

employer’s request in order to publicize 

job vacancies. The information is 

collected by One-Stop Career Centers 

and posted on electronic job banks. 20 

CFR 652.8(d)(5) specifies the one-year 

retention of information on work 

applications and job orders. For 

additional information, see related 

notice published in the Federal Register 

on January 5, 2010 (75 FR 450). 

Agency: Employment and Training 

Administration. 

Type of Review: Extension without 

change of a currently approved 

collection. 

Title of Collection: Benefit Rights and 

Experience Report. 

OMB Control Number: 1205–0177. 

Agency Form Number: ETA–218. 

Affected Public State, Local, or Tribal 

Governments. 

Total Estimated Number of 

Respondents: 53. 

Total Estimated Annual Burden 

Hours: 108. 

Total Estimated Annual Costs Burden 

(Operation and Maintenance): $0. 

Description: The Form ETA–218 

provides information used in solvency 

studies, in budgeting projections and for 

evaluation of adequacy of benefit 

formulas to analyze effects or proposed 

changes in state law. For additional 

information, see related notice 

published in the Federal Register on 

January 25, 2010 (75 FR 3927). 

Agency: Employment and Training 

Administration. 

Type of Review: Extension without 

change of a currently approved 

collection. 

Title of Collection: Transmittal of 

Unemployment Insurance Materials. 

OMB Control Number: 1205–0222. 

Agency Form Number: MA 8–7. 

Affected Public: State, Local, or Tribal 

Governments. 

Total Estimated Number of 

Respondents: 53. 

Total Estimated Annual Burden 

Hours: 11. 

Total Estimated Annual Costs Burden 

(Operation and Maintenance): $0. 

Description: Section 303(a)(6), Social 

Security Act, Public Law 74–271, (SSA), 

requires, as a condition of receiving 

administrative grants, that State law 

contain provision for the ‘‘making of 

such reports, in such form and 

containing such information, as the 

Secretary of Labor may from time to 

time require, and compliance with such 

provisions as the Secretary of Labor may 

from time to time find necessary to 

ensure the correctness and verification 

of such reports.’’ Departmental 

regulations at 20 CFR 601.3 in part 

implement this requirement by 

requiring the submission of ‘‘all relevant 

state materials, such as statutes, 

executive and administrative orders, 

legal opinions, rules, regulations, 

interpretations, court opinions, etc. 

***’’ Also, the regulations for the 

Unemployment Compensation (UC) for 

Federal Civilian Employees (UCFE) 

program at 20 CFR 609.1(d)(1) and for 

the UC for ex-service members (UCX) 

program at 20 CFR 614.1(d)(1) require 

submission of certain documents to 

assure that states are properly 

administering these programs. The 

Trade Adjustment Assistance (which 

includes Trade Readjustment 

Allowances) program (TAA/TRA) 

regulations provide similar 

requirements at 20 CFR 617.52(c)(1). 

The Form MA 8–7 is the mechanism 

for implementing these submittal 

requirements, the purpose of which is to 

provide the Secretary with sufficient 

information to determine if (a) 

employers in a state qualify for tax 

credits under the Federal 

Unemployment Tax Act; (b) the state 

meets the requirements for obtaining 

administrative grants under Title III, 

SSA; and (c) the state is fulfilling it 

obligations under Federal UC programs. 

For additional information, see related 

notice published in the Federal Register 

on January 25, 2010 (75 FR 3926). 

Darrin A. King, 

Departmental Clearance Officer. 

[FR Doc. 2010–10303 Filed 5–3–10; 8:45 am] 

BILLING CODE 4510–FW–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2010–0169] 

Biweekly Notice; Applications and 

Amendments to Facility Operating 

Licenses Involving No Significant 

Hazards Considerations 

I. Background 

Pursuant to section 189a. (2) of the 

Atomic Energy Act of 1954, as amended 

(the Act), the U.S. Nuclear Regulatory 

Commission (the Commission or NRC) 

is publishing this regular biweekly 

notice. The Act requires the 

Commission publish notice of any 

amendments issued, or proposed to be 

issued and grants the Commission the 

authority to issue and make 

immediately effective any amendment 

to an operating license upon a 

determination by the Commission that 

such amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

This biweekly notice includes all 

notices of amendments issued, or 

proposed to be issued from April 8, 

2010 to April 21, 2010. The last 

biweekly notice was published on April 

20, 2010 (75 FR 20627). 

Notice of Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The Commission has made a 

proposed determination that the 

following amendment requests involve 

no significant hazards consideration. 

Under the Commission’s regulations in 

Title 10 of the Code of Federal 

Regulations (10 CFR), Section 50.92, 

this means that operation of the facility 

in accordance with the proposed 

amendment would not (1) involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated; or (2) create the possibility of 

a new or different kind of accident from 

any accident previously evaluated; or 

(3) involve a significant reduction in a 

margin of safety. The basis for this 

proposed determination for each 

amendment request is shown below. 

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Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 23809 

The Commission is seeking public 

comments on this proposed 

determination. Any comments received 

within 30 days after the date of 

publication of this notice will be 

considered in making any final 

determination. 

Normally, the Commission will not 

issue the amendment until the 

expiration of 60 days after the date of 

publication of this notice. The 

Commission may issue the license 

amendment before expiration of the 60- 

day period provided that its final 

determination is that the amendment 

involves no significant hazards 

consideration. In addition, the 

Commission may issue the amendment 

prior to the expiration of the 30-day 

comment period should circumstances 

change during the 30-day comment 

period such that failure to act in a 

timely way would result, for example in 

derating or shutdown of the facility. 

Should the Commission take action 

prior to the expiration of either the 

comment period or the notice period, it 

will publish in the Federal Register a 

notice of issuance. Should the 

Commission make a final No Significant 

Hazards Consideration Determination, 

any hearing will take place after 

issuance. The Commission expects that 

the need to take this action will occur 

very infrequently. 

Written comments may be submitted 

by mail to the Chief, Rules, 

Announcements and Directives Branch 

(RADB), TWB–05–B01M, Division of 

Administrative Services, Office of 

Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, and should cite the publication 

date and page number of this Federal 

Register notice. Written comments may 

also be faxed to the RADB at 301–492– 

3446. Documents may be examined, 

and/or copied for a fee, at the NRC’s 

Public Document Room (PDR), located 

at One White Flint North, Public File 

Area O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. 

Within 60 days after the date of 

publication of this notice, any person(s) 

whose interest may be affected by this 

action may file a request for a hearing 

and a petition to intervene with respect 

to issuance of the amendment to the 

subject facility operating license. 

Requests for a hearing and a petition for 

leave to intervene shall be filed in 

accordance with the Commission’s 

‘‘Rules of Practice for Domestic 

Licensing Proceedings’’ in 10 CFR part 

2. Interested person(s) should consult a 

current copy of 10 CFR 2.309, which is 

available at the Commission’s PDR, 

located at One White Flint North, Public 

File Area O1F21, 11555 Rockville Pike 

(first floor), Rockville, Maryland. 

Publicly available records will be 

accessible from the Agencywide 

Documents Access and Management 

System’s (ADAMS) Public Electronic 

Reading Room on the Internet at the 

NRC Web site, http://www.nrc.gov/ 

reading-rm/doc-collections/cfr/. If a 

request for a hearing or petition for 

leave to intervene is filed by the above 

date, the Commission or a presiding 

officer designated by the Commission or 

by the Chief Administrative Judge of the 

Atomic Safety and Licensing Board 

Panel, will rule on the request and/or 

petition; and the Secretary or the Chief 

Administrative Judge of the Atomic 

Safety and Licensing Board will issue a 

notice of a hearing or an appropriate 

order. 

As required by 10 CFR 2.309, a 

petition for leave to intervene shall set 

forth with particularity the interest of 

the petitioner in the proceeding, and 

how that interest may be affected by the 

results of the proceeding. The petition 

should specifically explain the reasons 

why intervention should be permitted 

with particular reference to the 

following general requirements: (1) The 

name, address, and telephone number of 

the requestor or petitioner; (2) the 

nature of the requestor’s/petitioner’s 

right under the Act to be made a party 

to the proceeding; (3) the nature and 

extent of the requestor’s/petitioner’s 

property, financial, or other interest in 

the proceeding; and (4) the possible 

effect of any decision or order which 

may be entered in the proceeding on the 

requestor’s/petitioner’s interest. The 

petition must also identify the specific 

contentions which the requestor/ 

petitioner seeks to have litigated at the 

proceeding. 

Each contention must consist of a 

specific statement of the issue of law or 

fact to be raised or controverted. In 

addition, the requestor/petitioner shall 

provide a brief explanation of the bases 

for the contention and a concise 

statement of the alleged facts or expert 

opinion which support the contention 

and on which the requestor/petitioner 

intends to rely in proving the contention 

at the hearing. The requestor/petitioner 

must also provide references to those 

specific sources and documents of 

which the petitioner is aware and on 

which the requestor/petitioner intends 

to rely to establish those facts or expert 

opinion. The petition must include 

sufficient information to show that a 

genuine dispute exists with the 

applicant on a material issue of law or 

fact. Contentions shall be limited to 

matters within the scope of the 

amendment under consideration. The 

contention must be one which, if 

proven, would entitle the requestor/ 

petitioner to relief. A requestor/ 

petitioner who fails to satisfy these 

requirements with respect to at least one 

contention will not be permitted to 

participate as a party. 

Those permitted to intervene become 

parties to the proceeding, subject to any 

limitations in the order granting leave to 

intervene, and have the opportunity to 

participate fully in the conduct of the 

hearing. 

If a hearing is requested, the 

Commission will make a final 

determination on the issue of no 

significant hazards consideration. The 

final determination will serve to decide 

when the hearing is held. If the final 

determination is that the amendment 

request involves no significant hazards 

consideration, the Commission may 

issue the amendment and make it 

immediately effective, notwithstanding 

the request for a hearing. Any hearing 

held would take place after issuance of 

the amendment. If the final 

determination is that the amendment 

request involves a significant hazards 

consideration, any hearing held would 

take place before the issuance of any 

amendment. 

All documents filed in NRC 

adjudicatory proceedings, including a 

request for hearing, a petition for leave 

to intervene, any motion or other 

document filed in the proceeding prior 

to the submission of a request for 

hearing or petition to intervene, and 

documents filed by interested 

governmental entities participating 

under 10 CFR 2.315(c), must be filed in 

accordance with the NRC E-Filing rule 

(72 FR 49139, August 28, 2007). The EFiling process requires participants to 

submit and serve all adjudicatory 

documents over the Internet, or in some 

cases to mail copies on electronic 

storage media. Participants may not 

submit paper copies of their filings 

unless they seek an exemption in 

accordance with the procedures 

described below. 

To comply with the procedural 

requirements of E-Filing, at least ten 

(10) days prior to the filing deadline, the 

participant should contact the Office of 

the Secretary by e-mail at 

hearing.docket@nrc.gov, or by telephone 

at (301) 415–1677, to request (1) a 

digital ID certificate, which allows the 

participant (or its counsel or 

representative) to digitally sign 

documents and access the E-Submittal 

server for any proceeding in which it is 

participating; and (2) advise the 

Secretary that the participant will be 

submitting a request or petition for 

hearing (even in instances in which the 

participant, or its counsel or 

mstockstill on DSKH9S0YB1PROD with NOTICES

23810 Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 

representative, already holds an NRCissued digital ID certificate). Based upon 

this information, the Secretary will 

establish an electronic docket for the 

hearing in this proceeding if the 

Secretary has not already established an 

electronic docket. 

Information about applying for a 

digital ID certificate is available on 

NRC’s public Web site at http:// 

www.nrc.gov/site-help/e-submittals/ 

apply-certificates.html. System 

requirements for accessing the ESubmittal server are detailed in NRC’s 

‘‘Guidance for Electronic Submission,’’ 

which is available on the agency’s 

public Web site at http://www.nrc.gov/ 

site-help/e-submittals.html. Participants 

may attempt to use other software not 

listed on the Web site, but should note 

that the NRC’s E-Filing system does not 

support unlisted software, and the NRC 

Meta System Help Desk will not be able 

to offer assistance in using unlisted 

software. 

If a participant is electronically 

submitting a document to the NRC in 

accordance with the E-Filing rule, the 

participant must file the document 

using the NRC’s online, Web-based 

submission form. In order to serve 

documents through EIE, users will be 

required to install a Web browser plugin from the NRC Web site. Further 

information on the Web-based 

submission form, including the 

installation of the Web browser plug-in, 

is available on the NRC’s public Web 

site at http://www.nrc.gov/site-help/esubmittals.html. 

Once a participant has obtained a 

digital ID certificate and a docket has 

been created, the participant can then 

submit a request for hearing or petition 

for leave to intervene. Submissions 

should be in Portable Document Format 

(PDF) in accordance with NRC guidance 

available on the NRC public Web site at 

http://www.nrc.gov/site-help/esubmittals.html. A filing is considered 

complete at the time the documents are 

submitted through the NRC’s E-Filing 

system. To be timely, an electronic 

filing must be submitted to the E-Filing 

system no later than 11:59 p.m. Eastern 

Time on the due date. Upon receipt of 

a transmission, the E-Filing system 

time-stamps the document and sends 

the submitter an e-mail notice 

confirming receipt of the document. The 

E-Filing system also distributes an email notice that provides access to the 

document to the NRC Office of the 

General Counsel and any others who 

have advised the Office of the Secretary 

that they wish to participate in the 

proceeding, so that the filer need not 

serve the documents on those 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before a hearing request/ 

petition to intervene is filed so that they 

can obtain access to the document via 

the E-Filing system. 

A person filing electronically using 

the agency’s adjudicatory E-Filing 

system may seek assistance by 

contacting the NRC Meta System Help 

Desk through the ‘‘Contact Us’’ link 

located on the NRC Web site at http:// 

www.nrc.gov/site-help/esubmittals.html, by e-mail at 

MSHD.Resource@nrc.gov, or by a tollfree call at (866) 672–7640. The NRC 

Meta System Help Desk is available 

between 8 a.m. and 8 p.m., Eastern 

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing requesting authorization to 

continue to submit documents in paper 

format. Such filings must be submitted 

by: (1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, Attention: Rulemaking and 

Adjudications Staff; or (2) courier, 

express mail, or expedited delivery 

service to the Office of the Secretary, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland, 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

Filing is considered complete by firstclass mail as of the time of deposit in 

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. A presiding 

officer, having granted an exemption 

request from using E-Filing, may require 

a participant or party to use E-Filing if 

the presiding officer subsequently 

determines that the reason for granting 

the exemption from use of E-Filing no 

longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket which is 

available to the public at http:// 

ehd.nrc.gov/EHD_Proceeding/home.asp, 

unless excluded pursuant to an order of 

the Commission, or the presiding 

officer. Participants are requested not to 

include personal privacy information, 

such as social security numbers, home 

addresses, or home phone numbers in 

their filings, unless an NRC regulation 

or other law requires submission of such 

information. With respect to 

copyrighted works, except for limited 

excerpts that serve the purpose of the 

adjudicatory filings and would 

constitute a Fair Use application, 

participants are requested not to include 

copyrighted materials in their 

submission. 

Petitions for leave to intervene must 

be filed no later than 60 days from the 

date of publication of this notice. Nontimely filings will not be entertained 

absent a determination by the presiding 

officer that the petition or request 

should be granted or the contentions 

should be admitted, based on a 

balancing of the factors specified in 10 

CFR 2.309(c)(1)(i)–(viii). 

For further details with respect to this 

license amendment application, see the 

application for amendment which is 

available for public inspection at the 

Commission’s PDR, located at One 

White Flint North, Public File Area 

O1F21, 11555 Rockville Pike (first 

floor), Rockville, Maryland. Publicly 

available records will be accessible from 

the ADAMS Public Electronic Reading 

Room on the Internet at the NRC Web 

site, http://www.nrc.gov/reading-rm/ 

adams.html. Persons who do not have 

access to ADAMS or who encounter 

problems in accessing the documents 

located in ADAMS, should contact the 

NRC PDR Reference staff at 1–800–397– 

4209, 301–415–4737, or by e-mail to 

pdr.resource@nrc.gov. 

Calvert Cliffs Nuclear Power Plant, LLC, 

Docket Nos. 50–317 and 50–318, Calvert 

Cliffs Nuclear Power Plant, Unit Nos. 1 

and 2, Calvert County, Maryland 

Date of amendments request: 

November 23, 2009. 

Description of amendments request: 

The amendment would modify the 

licensing basis and the Technical 

Specifications by allowing for the 

transition from Westinghouse Turbo 

fuel to AREVA Advanced CE–14 High 

Thermal Performance (HTP) fuel in the 

Calvert Cliffs reactors. The licensee 

plans to refuel and operate with AREVA 

fuel beginning with the refueling outage 

in 2011 for Unit No. 2 and 2012 for Unit 

No. 1. The transition is planned to occur 

over three refueling cycles on each unit. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

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Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 23811 

No. 

The reactor fuel and the analyses 

associated with it are not accident initiators. 

The response of the fuel to an accident is 

analyzed using conservative techniques and 

the results are compared to approved 

acceptance criteria. These evaluation results 

will show that the fuel response to an 

accident is within approved acceptance 

criteria for both cores loaded with the new 

AREVA Advanced CE–14 HTP fuel and cores 

loaded with both AREVA and Westinghouse 

Turbo fuel. Therefore, the change in fuel 

design does not affect accident or transient 

initiation or consequences. 

The proposed change to the Safety Limit 

Technical Specification (2.1.1.2) does not 

require any physical change to any plant 

system, structure, or component. The change 

to establish the peak fuel centerline 

temperature as the safety limit is consistent 

with the Standard Review Plan (SRP) for 

ensuring that the fuel design limits are met. 

Operations and analysis will continue to be 

in compliance with Nuclear Regulatory 

Commission (NRC) regulations. The peak fuel 

centerline temperature is the basis for 

protecting the fuel and is consistent with the 

analogous wording for other pressurized 

water reactor (PWR) plants. Providing the 

peak fuel centerline melt temperature as the 

safety limit does not impact the initiation or 

the mitigation of an accident. 

The proposed change to remove the total 

planar radial peaking factor (FTXY, Technical 

Specification 3.2.2) is based on a 

methodology change. During and after the 

transition to AREVA Advanced CE–14 HTP 

fuel, the core analyses are performed using 

AREVA methodologies. These methodologies 

do not use the total planar radial peaking 

factor (FTXY) as an initial value in the 

accident analyses. The linear heat rate 

algorithm limits are provided by the total 

integrated radial peaking factor, azimuthal 

power tilt, and axial shape index. The linear 

heat rate is evaluated in accordance with 

NRC-approved methodology and meets 

acceptance criteria. The total planar radial 

peaking factor is not an accident initiator and 

does not play a role in accident mitigation. 

A number of other changes are also made to 

remove references to Technical Specification 

3.2.2 throughout the Technical 

Specifications. 

Topical reports have been reviewed and 

approved by the NRC for use in determining 

core operating limits. The core operating 

limits to be developed using the new 

methodologies will be established in 

accordance with the applicable limitations as 

documented in the appropriate NRC Safety 

Evaluation reports. The proposed change to 

add and remove various topical reports to 

Technical Specification 5.6.5 enables the use 

of appropriate methodologies to re-analyze 

certain events. The proposed methodologies 

will ensure that the plant continues to meet 

applicable design criteria and safety analysis 

acceptance criteria. 

The proposed change to the list of NRCapproved methodologies listed in Technical 

Specification 5.6.5 is administrative in nature 

and has no impact on any plant configuration 

or system performance relied upon to 

mitigate the consequences of an accident. 

The proposed change will update the listing 

of NRC-approved methodologies to remove 

methods no longer used and add new 

methods consistent with the transition to 

AREVA Advanced CE–14 HTP fuel. Changes 

to the calculated core operating limits may 

only be made using NRC-approved methods, 

must be consistent with all applicable safety 

analysis limits and are controlled by the 10 

CFR 50.59 process. The list of methodologies 

in the Technical Specifications does not 

impact either the initiation of an accident or 

the mitigation of its consequences. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different type of 

accident from any accident previously 

evaluated? 

No. 

Use of AREVA Advanced CE–14 HTP fuel 

in the Calvert Cliffs reactor cores is 

consistent with the current plant design 

bases and does not adversely affect any 

fission product barrier, nor does it alter the 

safety function of safety systems, structures, 

or components, or their roles in accident 

prevention or mitigation. The operational 

characteristics of AREVA Advanced CE–14 

HTP fuel are bounded by the safety analyses. 

The AREVA Advanced CE–14 HTP fuel 

design performs within fuel design limits and 

does not create the possibility of a new or 

different type of accident. 

The proposed change to the Safety Limit 

Technical Specification (2.1.1.2) does not 

require any physical change to any plant 

system, structure, or component, nor does it 

require any change in safety analysis 

methods or results. The existing analyses 

remain unchanged and do not affect any 

accident initiators that would create a new 

accident. 

The proposed change to remove the total 

planar radial peaking factor (FTXY, Technical 

Specification 3.2.2) is based on a change in 

analytical methods needed to support the 

physical fuel change. These methodologies 

do not use the total planar radial peaking 

factor (FTXY) as an initial value in the 

accident analysis. The total planar radial 

peaking factor does not play a role in 

accident mitigation and cannot create the 

possibility of a new or different kind of 

accident. A number of other changes are 

made to remove references to Technical 

Specification 3.2.2 throughout the Technical 

Specifications. 

The proposed change to the list of topical 

reports used to determine the core operating 

limits is administrative in nature and has no 

impact on any plant configuration or on 

system performance. It updates the list of 

NRC-approved topical reports used to 

develop the core operating limits. There is no 

change to the parameters within which the 

plant is normally operated. The possibility of 

a new or different accident is not created. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

No. 

Use of AREVA Advanced CE–14 HTP fuel 

is consistent with the current plant design 

bases and does not adversely affect any 

fission product barrier, nor does it alter the 

safety function of safety systems, structures, 

or components, or their roles in accident 

prevention or mitigation. The operational 

characteristics of AREVA Advanced CE–14 

HTP fuel are bounded by the safety analyses. 

The AREVA Advanced CE–14 HTP fuel 

design performs within fuel design limits. 

The proposed changes do not result in 

exceeding design basis limits. Therefore, all 

licensed safety margins are maintained. 

The proposed change to the Safety Limit 

Technical Specification (2.1.1.2) does not 

require any physical change to any plant 

system, structure, or component, nor does it 

require any change in safety analysis 

methods or results. Therefore, by changing 

the safety limit from peak linear heat rate to 

peak fuel centerline temperature, the margin 

as established in the current licensing basis 

remains unchanged. 

The proposed change to remove the total 

planar radial peaking factor (FTXY,Technical 

Specification 3.2.2) is based on a 

methodology change. The linear heat rate 

algorithm limits are provided by the total 

integrated radial peaking factor, azimuthal 

power tilt, and axial shape index. The linear 

heat rate is evaluated in accordance with 

NRC-approved methodology and meets 

acceptance criteria. Therefore, the margin as 

established for the linear heat rate remains 

unchanged. A number of other changes are 

made to remove references to Technical 

Specification 3.2.2 throughout the Technical 

Specifications. 

The proposed change to the list of topical 

reports does not amend the cycle specific 

parameters presently required by the 

Technical Specifications. The individual 

Technical Specifications continue to require 

operation of the plant within the bounds of 

the limits specified in the COLR [Core 

Operating Limits Report]. The proposed 

change to the list of analytical methods 

referenced in the COLR is administrative in 

nature and does not impact the margin of 

safety. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendments request involves no 

significant hazards consideration. 

Attorney for licensee: Carey Fleming, 

Sr. Counsel—Nuclear Generation, 

Constellation Generation Group, LLC, 

750 East Pratt Street, 17th floor, 

Baltimore, MD 21202. 

NRC Branch Chief: Nancy L. Salgado. 

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23812 Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 

Entergy Nuclear Operations, Inc., 

Docket No. 50–293, Pilgrim Nuclear 

Power Station, Plymouth County, 

Massachusetts 

Date of amendment request: March 

15, 2010. 

Description of amendment request: 

The proposed amendment would revise 

a Technical Specification (TS) to 

address the increased setpoints and 

setpoint tolerances for Safety Relief 

Valves (SRVs) and Spring Safety Valves 

(SSVs) and changes related to the 

replacement of four Target Rock twostage SRVs with more reliable threestage SRVs and two existing Dresser 

3.749 inch throat diameter SSVs with 

Dresser 4.956 inch diameter SSVs. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration which is presented below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change increases the 

allowable as-found SRV and SSV setpoint 

tolerance, determined by test after the valves 

have been removed from service, from ± 1% 

to ± 3%. The proposed change also increases 

the SRV and SSV setpoints. Analysis of these 

changes demonstrates that reactor pressure 

will be maintained below the applicable code 

overpressure limits. The proposed change 

increases the SSV discharge capacity due to 

its increased throat diameter. The proposed 

change does not alter the TS requirements for 

the number of SRVs and SSVs required to be 

operable, the allowable as-left lift setpoint 

tolerance, the testing frequency, or the 

manner in which the valves are operated. 

Consistent with current TS requirements, the 

proposed change continues to require that 

the safety valves be adjusted to within ± 1% 

of their nominal lift setpoints following 

testing. The proposed increase in the SRV 

and SSV setpoint complies with the ASME 

Boiler and Pressure Vessel (B&PV) Code 

(1965 Edition, including January 1966 

Addendum) for the pressure vessel, USAS 

Piping Code Section B31.1 for the steam 

space piping, and ASME Section III for the 

reactor coolant system recirculation piping. 

Since the proposed change does not alter the 

manner in which the valves are operated, 

there is no significant impact on the reactor 

operation. 

The proposed change does not involve a 

change to the safety function of the valves. 

The proposed TS revision involves no 

significant changes to the operation of any 

systems or components in normal or accident 

operating conditions. Therefore, these 

changes will not increase the probability of 

an accident previously evaluated. 

Since an SSV setpoint increase and 

setpoint tolerance will increase the SSV 

safety valve opening pressure and an increase 

in the SSV throat size will increase the SSV 

flow capacity, the SSV dynamic loads are 

expected to increase. Entergy has evaluated 

the SSV dynamic loads for the associated 

piping. All piping and structures were found 

to meet Code requirements. 

Since an SRV setpoint and the setpoint 

tolerance increase will increase the SRV 

valve opening pressure, the SRV discharge 

dynamic loads will increase. Entergy has 

evaluated the SRV dynamic load increases 

for the associated piping and torus 

submerged structures and the evaluation 

concluded that all piping and structures were 

found to meet Code requirements. 

The proposed revision to the HPCI [highpressure coolant injection] and RCIC [Reactor 

Core Isolation Cooling] pump operability 

determination surveillance follows the 

format of BWR Standard Technical 

Specification surveillance, and complies 

with in-service testing for pump operability 

determination in accordance with ASME OM 

Code requirement. 

Generic considerations related to the 

change in setpoints and setpoint tolerance 

were addressed in NEDC–31753P, ‘‘BWROG 

In-Service Pressure Relief Technical 

Specification Revision Licensing Topical 

Report,’’ and were reviewed and approved by 

the NRC in a safety evaluation dated March 

8, 1993. General Electric Hitachi Company 

(GEH) completed plant-specific analyses to 

assess the impact of increase in SRV and SSV 

setpoints and increase in the setpoint 

tolerance from ± 1% to ± 3%. The impact of 

the increases in the SRV and SSV setpoints 

and increases in the setpoint tolerances, as 

addressed in this analysis, included vessel 

overpressure, Updated Final Safety Analysis 

Report (UFSAR) Chapter 14 events, ATWS 

[Anticipated Transient Without Scram], Loss 

of Coolant Accident (LOCA), containment 

response and dynamic loads, high-pressure 

systems performance, operating mode and 

equipment out of service. The proposed 

change is supported by GEH analysis of 

events that credit the SRVs and SSVs. 

The plant specific evaluations, required by 

the NRC’s safety evaluation and performed to 

support this proposed change, demonstrate 

that there is no change to the design core 

thermal limits and adequate margin to the 

reactor coolant system pressure limits exists. 

These analyses also demonstrate that 

operation of Core Standby Cooling Systems 

(CSCS) is not adversely affected and the 

containment response following a LOCA is 

acceptable. The plant systems associated 

with these proposed changes are capable of 

meeting applicable design basis requirements 

and retain the capability to mitigate the 

consequences of accidents described in the 

UFSAR. Therefore, these changes do not 

involve an increase in the consequences of an 

accident previously evaluated. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed change increases the 

allowable as-found lift setpoint tolerance for 

the Pilgrim SRV and SSV valves. The 

proposed change to increase the tolerance 

was developed in accordance with the 

provisions contained in the NRC safety 

evaluation for NEDC–31753P. SRVs and 

SSVs installed in the plant following testing 

will continue to meet the current tolerance 

acceptance criteria of ± 1% of the nominal 

setpoint. The proposed change does not 

affect the manner in which the overpressure 

protection system is operated; therefore, 

there are no new failure mechanisms for the 

overpressure protection system. 

The proposed changes do not change the 

safety function of the SRVs and SSVs, or 

HPCI and RCIC systems. There is no 

alteration to the parameters within which the 

plant is normally operated. The increase in 

SRV and SSV setpoints, setpoint tolerance, 

and increased SSV discharge capacity are not 

precursors to new or different kinds of 

accidents and do not initiate new or different 

kinds of accidents. The impact of these 

changes have been analyzed and found to be 

acceptable within the design limits and plant 

operating procedures. 

As a result, no new failure modes are being 

introduced. Therefore, the proposed change 

does not create the possibility of a new or 

different kind of accident from any 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The margin of safety is established through 

the design of the plant structures, systems, 

and components, the parameters within 

which the plant is operated and the 

establishment of the setpoints for the 

actuation of equipment relied upon to 

respond to an event. The proposed change 

modifies the setpoints at which protective 

actions are initiated, and [* * *] does not 

change the requirements governing operation 

or availability of safety equipment assumed 

to operate to preserve the margin of safety. 

Establishment of the ± 3% SRV and SSV 

setpoint tolerance limit does not adversely 

affect the operation of any safety-related 

component or equipment. Evaluations 

performed in accordance with the NRC safety 

evaluation for NEDC–31753P have concluded 

that all design limits will continue to be met. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. William C. 

Dennis, Assistant General Counsel, 

Entergy Nuclear Operations, Inc., 400 

Hamilton Avenue, White Plains, NY 

10601. 

NRC Branch Chief: Nancy Salgado. 

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Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 23813 

Entergy Operations, Inc., Docket No. 50– 

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: February 

22, 2010. 

Description of amendment request: 

The proposed amendment will modify 

Technical Specification (TS) 3/4.9.4, 

‘‘Containment Building Penetrations,’’ to 

allow alternative means of penetration 

closure during Core Alterations or 

irradiated fuel movement while in 

refueling operations. Additional 

improvements to the TS are also being 

proposed, as well as the elimination of 

TS 3/4.9.9, ‘‘Containment Purge Valve 

Isolation System.’’ The proposed 

changes are consistent with Revision 3 

of NUREG–1432, ‘‘Standard Technical 

Specifications Combustion Engineering 

Plants.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

TS 3/4.9.4 currently allows containment 

penetration flow paths to be open during 

Core Alterations or movement of irradiated 

fuel within containment under specific 

administrative controls. The proposed 

change would allow additional approved 

methods for ensuring positive penetration 

closure. The fuel handling accident (FHA) 

radiological analysis does not take credit for 

containment isolation or filtration. Therefore, 

the time required to close any open 

penetrations does not affect the radiological 

analysis dose calculations and the proposed 

change does not involve a significant 

increase in the consequences of an accident 

previously evaluated. The administrative 

controls for containment penetration closure 

are conservative even though not required by 

the accident analysis. 

The proposed revision only provides 

alternate methods of penetration closure and 

does not alter any plant equipment where the 

probability of an accident would be 

increased. The incorporation of purge valve 

isolation surveillance requirements for 

assuring purge valve Operability has no effect 

on the probability or consequences of the 

analyzed accidents. 

Therefore, this change does not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Alternative methods of providing 

penetration closure do not create accident 

initiators and do not represent a significant 

change in the configuration of the plant. The 

proposed allowance to secure containment 

penetrations during refueling operations will 

not adversely effect plant safety functions or 

equipment operating practices such that a 

new or different accident could be created. 

Therefore, the proposed change does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

TS Limiting Condition for Operation (LCO) 

3.9.4 closure requirements for containment 

penetrations ensure that the consequences of 

a postulated FHA inside containment during 

Core Alterations or fuel handling activities 

are minimized. The LCO establishes 

containment closure requirements, which 

limit the potential escape paths for fission 

products by ensuring that there is at least one 

barrier to the release of radioactive material. 

The proposed change to allow alternate 

methods of reaching containment penetration 

closure during Core Alterations or fuel 

movement does not affect the expected dose 

consequences of a FHA since it does not 

credit containment building closure. The 

proposed administrative controls provide 

assurance that prompt closure of the 

penetration flow paths will be accomplished 

in the event of a FHA inside containment 

thus minimizing the transmission of 

radioactive material from the containment to 

the outside environment. The incorporation 

of purge valve isolation surveillance 

requirements does not reduce any margins of 

safety. 

Therefore, the proposed change does not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Associate General Council— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC Branch Chief: Michael T. 

Markley. 

Entergy Operations, Inc., Docket No. 50– 

382, Waterford Steam Electric Station, 

Unit 3, St. Charles Parish, Louisiana 

Date of amendment request: February 

24, 2010. 

Description of amendment request: 

The proposed amendment deletes 

Operating License Condition 2.C.14 

(Fuel Movement in the Fuel Handling 

Building) due to electing to comply with 

Section 50.68, ‘‘Criticality accident 

requirements,’’ of Title 10 of the Code of 

Federal Regulations (10 CFR). The 

Operating License Condition 2.C.14, ‘‘no 

more than one fuel assembly shall be 

out of its shipping container or storage 

location at a given time,’’ was one basis 

for the exemption from the criticality 

alarm system requirements of 10 CFR 

70.24. The criticality accident 

requirements can be met either by 

complying with 10 CFR 70.24 or 10 CFR 

50.68 requirements. The 10 CFR 50.68 

criteria are now being used; therefore, 

Operating License Condition 2.C.14 is 

no longer applicable. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed amendment deletes 

Operating License Condition 2.C.14 (Fuel 

Movement in the Fuel Handling Building) 

due to electing to comply with 10 CFR 50.68 

requirements. 

The proposed changes will not alter the 

configuration of the storage racks or their 

environment. The fuel racks will not be 

operated outside of their design limits, and 

no additional loads will be imposed on them. 

Therefore, these changes will not affect fuel 

storage rack performance or reliability. No 

new equipment will be introduced into the 

plant. The accuracies and response 

characteristics of existing instrumentation 

will not be modified. The proposed changes 

will not require, or result in, a change in 

safety system operation, and will not affect 

any system interface with the fuel storage 

racks. Fuel assembly placement will continue 

to be controlled in accordance with approved 

fuel handling procedures. All the 

requirements of 10 CFR 50.68 continue to be 

met which ensures no significant increase in 

the probability or consequences of an 

accident previously evaluated. 

The proposed changes will not affect any 

barrier that mitigates dose to the public, and 

will not result in a new release pathway 

being created. The functions of equipment 

designed to control the release of radioactive 

material will not be impacted, and no 

mitigating actions described or assumed for 

an accident in the UFSAR [Updated Final 

Safety Analysis Report] will be altered or 

prevented. No assumptions previously made 

in evaluating the consequences of an 

accident will need to be modified. Onsite 

dose will not be increased, so the access of 

plant personnel to vital areas of the plant will 

not be restricted, and mitigating actions will 

not be impeded. 

Therefore, it is concluded that the 

proposed changes do not significantly 

increase either the probability or 

consequences of any accident previously 

evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

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23814 Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 

accident from any accident previously 

evaluated? 

Response: No. 

The proposed amendment deletes 

Operating License Condition 2.C.14 (Fuel 

Movement in the Fuel Handling Building) 

due to electing to comply with 10 CFR 50.68 

requirements. 

10 CFR 50.68(b)(1) provides the 

requirements to ensure that plant procedures 

shall prohibit the handling and storage at any 

one time of more fuel assemblies than have 

been determined to be safely subcritical 

under the most adverse moderation 

conditions feasible by unborated water. By 

meeting this criteria, the removal of 

Operating License Condition 2.C.14 will not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

Therefore, it is concluded that the 

proposed changes do not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The proposed amendment deletes 

Operating License Condition 2.C.14 (Fuel 

Movement in the Fuel Handling Building) 

due to electing to comply with 10 CFR 50.68 

requirements. 

10 CFR 50.68(b)(1) provides similar 

requirements as that contained in Operating 

License Condition 2.C.14. The NRC has 

approved the [Waterford Steam Electric 

Station, Unit 3] use of 10 CFR 50.68 criteria. 

By meeting the 10 CFR 50.68(b)(1) 

requirements, there will not be a significant 

reduction in a margin of safety. 

Therefore, it is concluded that the 

proposed changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Joseph A. 

Aluise, Associate General Council— 

Nuclear, Entergy Services, Inc., 639 

Loyola Avenue, New Orleans, Louisiana 

70113. 

NRC Branch Chief: Michael T. 

Markley. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois 

Date of amendment request: February 

15, 2010. 

Description of amendment request: 

The proposed amendment would 

relocate selected Surveillance 

Requirement frequencies from the 

Clinton Power Station, Unit No. 1 

(Clinton) Technical Specifications (TSs) 

to a licensee-controlled program. This 

change is based on the NRC-approved 

Industry Technical Specifications Task 

Force (TSTF) change TSTF–425, 

‘‘Relocate Surveillance Frequencies to 

Licensee Control—Risk Informed 

Technical Specification Task Force 

(RITSTF) Initiative 5b,’’ Revision 3, 

(Agencywide Documents Access and 

Management System (ADAMS) 

Accession Package No. ML090850642). 

Plant-specific deviations from TSTF– 

425 are proposed to accommodate 

differences between the Clinton TSs and 

the model TSs originally used to 

develop TSTF–425. 

The Nuclear Regulatory Commission 

(NRC) staff issued a Notice of 

Availability for TSTF–425 in the 

Federal Register on July 6, 2009 (74 FR 

31996). The notice included a model 

safety evaluation (SE) and a model no 

significant hazards consideration 

(NSHC) determination. In its application 

dated February 15, 2010 (ADAMS 

Accession No. ML100470787), the 

licensee affirmed the applicability of the 

model NSHC determination which is 

presented below. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), an 

analysis of the issue of NSHC is 

presented below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. The proposed change 

relocates the specified frequencies for 

periodic surveillance requirements to 

licensee control under a new Surveillance 

Frequency Control Program. Surveillance 

frequencies are not an initiator to any 

accident previously evaluated. As a result, 

the probability of any accident previously 

evaluated is not significantly increased. The 

systems and components required by the 

technical specifications for which the 

surveillance frequencies are relocated are 

still required to be operable, meet the 

acceptance criteria for the surveillance 

requirements, and be capable of performing 

any mitigation function assumed in the 

accident analysis. As a result, the 

consequences of any accident previously 

evaluated are not significantly increased. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. No new or different 

accidents result from utilizing the proposed 

change. The changes do not involve a 

physical alteration of the plant (i.e., no new 

or different type of equipment will be 

installed) or change in the methods 

governing normal plant operation. In 

addition, the changes do not impose any new 

or different requirements. The changes do not 

alter assumptions made in the safety 

analysis. The proposed changes are 

consistent with the safety analysis 

assumptions and current plant operating 

practice. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Does the proposed change involve a 

significant reduction in the margin of safety? 

Response: No. The design, operation, 

testing methods, and acceptance criteria for 

systems, structures, and components (SSCs), 

specified in applicable codes and standards 

(or alternatives approved for use by the NRC) 

will continue to be met as described in the 

plant licensing basis (including the final 

safety analysis report and bases to TS), since 

these are not affected by changes to the 

surveillance frequencies. Similarly, there is 

no impact to safety analysis acceptance 

criteria as described in the plant licensing 

basis. To evaluate a change in the relocated 

surveillance frequency, Exelon will perform 

a probabilistic risk evaluation using the 

guidance contained in NRC approved NEI 

04–01, Rev. 1. The methodology provides 

reasonable acceptance guidelines and 

methods for evaluating the risk increase of 

proposed changes to surveillance frequencies 

consistent with Regulatory Guide 1.177 [An 

Approach for Plant-Specific, Risk-Informed 

Decision-making: Technical Specifications]. 

Therefore, the proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Stephen J. 

Campbell. 

Exelon Generation Company, LLC, 

Docket No. 50–461, Clinton Power 

Station, Unit No. 1, DeWitt County, 

Illinois 

Date of amendment request: March 3, 

2010. 

Description of amendment request: 

The proposed amendment revises 

Technical Specification (TS) 3.1.7, 

‘‘Standby Liquid Control (SLC) System,’’ 

to extend the completion time (CT) for 

Condition B (i.e., ‘‘Two SLC subsystems 

inoperable’’) from 8 hours to 72 hours. 

Basis for proposed no significant 

hazards consideration: As required by 

10 CFR 50.91(a), an analysis of the issue 

of no significant hazards consideration 

is presented below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

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Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 23815 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed amendment revises 

Technical Specification (TS) 3.1.7, ‘‘Standby 

Liquid Control (SLC) System,’’ to extend the 

completion time (CT) for Condition B (i.e., 

‘‘Two SLC subsystems inoperable.’’) from 

eight hours to 72 hours. 

The proposed change is based on a riskinformed evaluation performed in 

accordance with Regulatory Guides (RG) 

1.174, ‘‘An Approach for Using Probabilistic 

Risk Assessment in Risk-Informed Decisions 

On Plant-Specific Changes to the Licensing 

Basis,’’ and RG 1.I77, ‘‘An Approach for 

Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications.’’ 

The proposed amendment modifies an 

existing CT for a dual-train SLC system 

inoperability. The condition evaluated, the 

action requirements, and the associated CT 

do not impact any initiating conditions for 

any accident previously evaluated. 

The proposed amendment does not 

increase postulated frequencies or the 

analyzed consequences of an Anticipated 

Transient Without Scram (ATWS). 

Requirements associated with 10 CFR 50.62 

will continue to be met. In addition, the 

proposed amendment does not increase 

postulated frequencies or the analyzed 

consequences or a large-break loss-of-coolant 

accident for which the SLC system will be 

used for pH control. The extended CT 

provides additional time to implement 

actions in response to a dual-train SLC 

system inoperability, while also minimizing 

the risk associated with continued operation. 

Therefore, the proposed change does not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

The proposed amendment revises TS 3.1.7 

to extend the CT for Condition B from eight 

hours to 72 hours. The proposed amendment 

does not involve any change to plant 

equipment or system design functions. This 

proposed TS amendment does not change the 

design function of the SLC system and does 

not affect the system’s ability to perform its 

design function. The SLC system provides a 

method to bring the reactor, at any time in 

a fuel cycle, from full power and minimum 

control rod inventory to a subcritical 

condition with the reactor in the most 

reactive xenon free state without taking 

credit for control rod movement. Required 

actions and surveillance requirements are 

sufficient to ensure that the SLC system 

functions are maintained. No new accident 

initiators are introduced by this amendment. 

Therefore, the proposed amendment does not 

create the possibility of a new or different 

kind of accident from any previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed amendment revises TS 3.1.7 

to extend the CT for Condition B from eight 

hours to 72 hours. The proposed amendment 

does not involve any change to plant 

equipment or system design functions. The 

margin of safety is established through the 

design of the plant structures, systems, and 

components, the parameters within which 

the plant is operated, and the setpoints for 

the actuation of equipment relied upon to 

respond to an event. 

The proposed amendment does not modify 

the condition or point at which SLC is 

initiated, nor does it affect the system’s 

ability to perform its design function. In 

addition, the proposed change complies with 

the intent of the defense-in-depth philosophy 

and the principle that sufficient safety 

margins are maintained, consistent with RG 

1.177 requirements (i.e., Section C, 

‘‘Regulatory Position,’’ paragraph 2.2 

‘‘Traditional Engineering considerations’’). 

Based on the above analysis, EGC 

concludes that the proposed amendment 

presents no significant hazards consideration 

under the standards set forth in 10 CFR 

50.92(c), and, accordingly, a finding of ‘‘no 

significant hazards consideration’’ is justified. 

The NRC staff has reviewed the 

analysis adopted by the licensee and, 

based on this review, it appears that the 

three standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Attorney for licensee: Mr. Bradley J. 

Fewell, Associate General Counsel, 

Exelon Generation Company, LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Stephen J. 

Campbell. 

Exelon Generation Company, LLC, and 

PSEG Nuclear, LLC, Docket Nos. 50–277 

and 50–278, Peach Bottom Atomic 

Power Station (PBAPS), Units 2 and 3, 

York and Lancaster Counties, 

Pennsylvania 

Date of amendment request: August 

31, 2009. 

Description of amendment request: 

The proposed amendment would 

modify the PBAPS Technical 

Specifications (TS) by relocating 

specific surveillance frequencies to a 

licensee-controlled program with the 

implementation of Nuclear Energy 

Institute (NEI) 04–10, ‘‘Risk-Informed 

Technical Specifications Initiative 5b, 

Risk-Informed Method for Control of 

Surveillance Frequencies.’’ 

Additionally, the change would add a 

new program, the Surveillance 

Frequency Control Program, to TS 

Section 5, Administrative Controls. The 

changes are based on NRC-approved 

Industry Technical Specifications Task 

Force (TSTF) Traveler 425, Revision 3, 

‘‘Relocate Surveillance Frequencies to 

Licensee Control—Risk Informed 

Technical Specification Task Force 

Initiative 5b,’’ with optional changes and 

variations as described in Attachment 1, 

Section 2.2 of the licensee’s submittal 

dated August 31, 2009. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Do the proposed changes involve a 

significant increase in the probability or 

consequences of any accident previously 

evaluated? 

Response: No. 

The proposed changes relocate the 

specified frequencies for periodic 

surveillance requirements to licensee control 

under a new Surveillance Frequency Control 

Program [SFCP]. Surveillance frequencies are 

not an initiator to any accident previously 

evaluated. As a result, the probability of any 

accident previously evaluated is not 

significantly increased. The systems and 

components required by the technical 

specifications for which the surveillance 

frequencies are relocated are still required to 

be operable, meet the acceptance criteria for 

the surveillance requirements, and be 

capable of performing any mitigation 

function assumed in the accident analysis. 

As a result, the consequences of any accident 

previously evaluated are not significantly 

increased. 

Therefore, the proposed changes do not 

involve a significant increase in the 

probability or consequences of an accident 

previously evaluated. 

2. Do the proposed changes create the 

possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

No new or different accidents result from 

utilizing the proposed changes. The changes 

do not involve a physical alteration of the 

plant (i.e., no new or different type of 

equipment will be installed) or a change in 

the methods governing normal plant 

operation. In addition, the changes do not 

impose any new or different requirements. 

The changes do not alter assumptions made 

in the safety analysis. The proposed changes 

are consistent with the safety analysis 

assumptions and current plant operating 

practice. 

Therefore, the proposed changes do not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. Do the proposed changes involve a 

significant reduction in the margin of safety? 

Response: No. 

[* * * T]here is no impact to safety 

analysis acceptance criteria as described in 

the plant licensing basis. To evaluate a 

change in the relocated surveillance 

frequency, Exelon will perform a 

probabilistic risk evaluation using the 

guidance contained in NRC approved NEI 

04–10, Rev. 1 in accordance with the TS 

SFCP. NEI 04–10, Rev. 1, methodology 

provides reasonable acceptance guidelines 

and methods for evaluating the risk increase 

of proposed changes to surveillance 

frequencies consistent with Regulatory Guide 

1.177. Therefore, the proposed changes do 

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23816 Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 

not involve a significant reduction in a 

margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, and with the changes noted 

above, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves NSHC. 

Attorney for licensee: Mr. J. Bradley 

Fewell, Associate General Counsel, 

Exelon Generation Company LLC, 4300 

Winfield Road, Warrenville, IL 60555. 

NRC Branch Chief: Harold K. 

Chernoff. 

FPL Energy Seabrook, LLC Docket No. 

50–443, Seabrook Station, Unit No. 1, 

Rockingham County, New Hampshire 

Date of amendment request: March 

16, 2010. 

Description of amendment request: 

The proposed changes would revise the 

Seabrook Technical Specifications 

requirement that the Operations 

Manager shall have held a senior reactor 

operator license for the Seabrook Station 

prior to assuming the Operations 

Manager position. Specifically, the 

proposed change would require the 

Operations Manager to meet one of the 

following: (1) Hold a senior operator 

license; (2) have held a senior operator 

license for a similar unit; or (3) have 

been certified for equivalent senior 

operator knowledge. In its application 

dated March 16, 2010, the licensee 

concluded that the no significant 

hazards consideration (NSHC) 

determination presented in the notice is 

applicable to Seabrook Station. 

Basis for proposed NSHC 

determination: As required by 10 CFR 

50.91(a), the licensee has provided its 

analysis of the issue of NSHC, which is 

presented below: 

1. The proposed change does not involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated. 

[The requested change would only affect 

the qualification requirements for the 

Operations Manager Position]. The proposed 

change does not impact the configuration or 

function of plant structures, systems, or 

components (SSCs) or the manner in which 

SSCs are operated, maintained, modified, 

tested, or inspected. No actual facility 

equipment or accident analyses will be 

affected by the proposed changes. Therefore, 

this request has no [significant] impact on the 

probability or consequences of an accident 

previously evaluated. 

2. The proposed change does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

[The requested change would only affect 

the qualification requirements for the 

Operations Manager Position]. The proposed 

change does not alter the plant configuration, 

require new plant equipment to be installed, 

alter accident analysis assumptions, add any 

initiators, or affect the function of plant 

systems or the manner in which systems are 

operated, maintained, modified, tested, or 

inspected. Therefore, this request does not 

create the possibility of a new or different 

kind of accident from any accident 

previously evaluated. 

3. The proposed change does not involve 

a significant reduction in a margin of safety. 

Margin of safety is associated with 

confidence in the ability of the fission 

product barriers (i.e., fuel cladding, reactor 

coolant system pressure boundary, and 

containment structure) to limit the level of 

radiation dose to the public. [The requested 

change would only affect the qualification 

requirements for the Operations Manager 

Position]. No actual plant equipment or 

accident analyses will be affected by the 

proposed changes. Additionally, the 

proposed changes will not relax any criteria 

used to establish safety limits, will not relax 

any safety system settings, and will not relax 

the bases for any limiting conditions for 

operation. The safety analysis acceptance 

criteria are not affected by this change. The 

proposed change will not result in plant 

operation in a configuration outside the 

design basis. The proposed change does not 

adversely affect systems that respond to 

safely shutdown the plant and to maintain 

the plant in a safe shutdown condition. 

Therefore, these proposed changes do not 

involve a significant reduction in a margin of 

safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, and with the changes noted 

above, it appears that the three 

standards of 50.92(c) are satisfied. 

Therefore, the NRC staff proposes to 

determine that the amendment request 

involves NSHC. 

Attorney for licensee: M.S. Ross, 

Florida Power & Light Company, P.O. 

Box 14000, Juno Beach, FL 33408–0420. 

NRC Branch Chief: Harold K. 

Chernoff. 

Northern States Power Company— 

Minnesota, Docket Nos. 50–282 and 50– 

306, Prairie Island Nuclear Generating 

Plant, Units 1 and 2 (PINGP), Goodhue 

County, Minnesota 

Date of amendment request: 

November 24, 2009. 

Description of amendment request: 

The proposed amendments would make 

changes to Technical Specification (TS) 

Section 4.2.1, Fuel Assemblies, and TS 

Section 5.6.5, Core Operating Limit 

Report, by revising the TS to allow the 

use of Optimized ZIRLOTM fuel rod 

cladding material. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed change involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

Westinghouse Electric Company, LLC 

(Westinghouse) topical report WCAP–12610– 

P–A and CENPD–404–P–A, Addendum 1–A 

‘‘Optimized ZIRLOTM’’, July 2006, provides 

the details and results of material testing of 

Optimized ZIRLOTM compared to standard 

ZIRLOTM as well as the material properties 

to be used in various models and 

methodologies when analyzing Optimized 

ZIRLOTM. The Nuclear Regulatory 

Commission (NRC) has allowed use of 

Optimized ZIRLOTM fuel cladding material 

in Westinghouse fueled reactors provided 

that licensees ensure compliance with the 

conditions and limitations set forth in the 

NRC Safety Evaluation (SE) for the topical 

report. By satisfying the conditions and 

limitations of the NRC SE through completed 

actions and its approved reload safety 

evaluation process, the licensee ensures that 

the effects of Optimized ZIRLOTM on PINGP 

core performance are evaluated and that the 

probability or consequences of previouslyevaluated accidents are not increased. 

Therefore, the proposed change of adding 

a cladding material does not result in an 

increase to the probability or consequences of 

an accident previously evaluated. 

2. Does the proposed change create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

Material properties of this fuel design have 

been evaluated in Westinghouse topical 

report WCAP–12610–P–A and CENPD–404– 

P–A, Addendum 1–A ‘‘Optimized ZIRLOTM’’ 

July 2006. That report provides the details 

and results of material testing of Optimized 

ZIRLOTM compared to standard ZIRLOTM as 

well as the material properties to be used in 

various models and methodologies when 

analyzing Optimized ZIRLOTM. Neither that 

topical report nor the associated NRC SE 

identifies the possibility of a new or different 

kind of accident resulting from this change 

for generic application in Westinghouse 

reactors. As demonstrated in that topical 

report and stated in the NRC SE, there is 

reasonable assurance that under both normal 

and accident conditions, the Optimized 

ZIRLOTM fuel cladding will be able to safely 

operate and comply with NRC regulations. 

By satisfying the conditions and limitations 

of the NRC SE by virtue of its completed 

actions and its approved reload safety 

evaluation process, the licensee ensures that 

the effects of Optimized ZIRLOTM are 

evaluated and will not create the possibility 

of a new or different kind of accident. 

Assurance that the possibility of new or 

different type of accidents will not be created 

on a site-specific basis is inherent to the 

reload safety evaluation process approved for 

use at the PINGP. Site specific evaluation of 

the PINGP core designs with Optimized 

ZIRLOTM will be performed 

programmatically and necessarily by the 

approved reload safety evaluation process. 

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Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 23817 

Therefore, the proposed change of adding 

a cladding material does not create the 

possibility of a new or different kind of 

accident from any accident previously 

evaluated. 

3. Does the proposed change involve a 

significant reduction in a margin of safety? 

Response: No. 

The cladding material used in the fuel rods 

is designed and tested to prevent excessive 

fuel temperatures, excessive internal rod gas 

pressure due to fission gas releases, and 

excessive cladding stresses and strains. 

Optimized ZIRLOTM was developed to meet 

these needs and provides a reduced corrosion 

rate while maintaining the benefits of 

mechanical strength and resistance to 

accelerated corrosion from abnormal 

chemistry conditions. Westinghouse topical 

report WCAP–12610–P–A and CENPD–404– 

P–A, Addendum 1–A ‘‘Optimized ZIRLOTM, 

July 2006, provides the details and results of 

material testing of Optimized ZIRLOTM

compared to standard ZIRLOTM as well as the 

material properties to be used in various 

models and methodologies when analyzing 

Optimized ZIRLOTM. The NRC has allowed 

use of Optimized ZIRLOTM fuel cladding 

material detailed within this topical report as 

detailed within their SE. Therefore, the 

change in material does not result in a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Peter M. Glass, 

Assistant General Counsel, Xcel Energy 

Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Northern States Power Company— 

Minnesota, Docket Nos. 50–282 and 50– 

306, Prairie Island Nuclear Generating 

Plant, Units 1 and 2, Goodhue County, 

Minnesota 

Date of amendment request: January 

27, 2010. 

Description of amendment request: 

The proposed amendments would make 

changes to the Technical Specifications 

(TS) to revise TS 3.8.3, ‘‘Diesel Fuel Oil’’. 

The amendments would revise the 

diesel fuel oil (DFO) storage volumes 

applicable to Unit 1 in TS 3.8.3 

Condition statements A and D, and 

increase the Unit 1 DFO supply required 

by surveillance requirement 3.8.3.1. The 

amendments would clarify wording in 

TS 3.8.3 Condition B statement which 

applies to both units. 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to increase the emergency diesel generator 

fuel oil storage volumes specified in the 

Technical Specification Condition statements 

and Surveillance Requirements. Also a word 

was added to a Condition statement to clarify 

its meaning. 

The emergency diesel generators and their 

supporting diesel fuel oil storage systems are 

not accident initiators and therefore the 

proposed fuel oil storage volume increases do 

not involve an increase in the probability of 

an accident. 

The proposed increased diesel fuel oil 

storage volumes provide sufficient volumes 

to maintain the current licensing basis for 

emergency diesel generator operation. Thus 

the proposed fuel oil storage volume 

increases do not involve a significant 

increase in the consequences of an accident. 

The proposed Technical Specification 

Condition statement wording clarification is 

administrative and thus does not involve an 

increase in the probability of an accident or 

an increase in the consequences of an 

accident. 

Therefore, the proposed Technical 

Specification changes do not involve a 

significant increase in the probability or 

consequences of an accident previously 

evaluated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any accident previously 

evaluated? 

Response: No. 

This license amendment request proposes 

to increase the emergency diesel generator 

fuel oil storage volumes specified in the 

Technical Specification Condition statements 

and Surveillance Requirements. Also a word 

was added to a Condition statement to clarify 

its meaning. 

The proposed Technical Specification 

changes which increase emergency diesel 

generator fuel oil storage volumes do not 

change any system operations or 

maintenance activities. The changes do not 

involve physical alteration of the plant, that 

is, no new or different type of equipment will 

be installed. The changes do not alter 

assumptions made in the safety analyses but 

ensures that the diesel generators operate as 

assumed in the accident analyses. These 

changes do not create new failure modes or 

mechanisms which are not identifiable 

during testing and no new accident 

precursors are generated. 

The proposed Technical Specification 

Condition statement wording clarification is 

administrative and thus does not create the 

possibility of a new or different kind of 

accident. 

Therefore, the proposed Technical 

Specification changes do not create the 

possibility of a new or different kind of 

accident from any previously evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

This license amendment request proposes 

to increase the emergency diesel generator 

fuel oil storage volumes specified in the 

Technical Specification Condition statements 

and Surveillance Requirements. Also a word 

was added to a Condition statement to clarify 

its meaning. 

Since this license amendment proposes 

Technical Specification changes which 

increase the required fuel oil storage 

volumes, margins of safety are increased and 

thus no margin of safety is reduced as part 

of this change. 

The proposed Technical Specification 

Condition statement wording clarification is 

administrative and thus does not involve a 

significant reduction in a margin of safety. 

Therefore, the proposed Technical 

Specification changes do not involve a 

significant reduction in a margin of safety. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment requests involve no 

significant hazards consideration. 

Attorney for licensee: Peter M. Glass, 

Assistant General Counsel, Xcel Energy 

Services, Inc., 414 Nicollet Mall, 

Minneapolis, MN 55401. 

NRC Branch Chief: Robert J. 

Pascarelli. 

Southern Nuclear Operating Company, 

Inc., Docket Nos. 50–424 and 50–425, 

Vogtle Electric Generating Plant, Units 1 

and 2, Burke County, Georgia 

Date of amendment request: February 

2, 2010. 

Description of amendment request: 

The proposed amendments would 

revise the verification requirements for 

the Reactor Trip System 

Instrumentation. Specifically, the 

amendment proposes the addition to 

Table 3.3.1–1 of a response time 

measurement for the verification of the 

Power Range Neutron High Positive 

Rate Trip (PFRT) function as 

recommended by Westinghouse Nuclear 

Safety Advisory Letter (NSAL–09–01) 

‘‘Rod Withdrawal at Power Analysis for 

Reactor Coolant System Overpressure.’’ 

Basis for proposed no significant 

hazards consideration determination: 

As required by 10 CFR 50.91(a), the 

licensee has provided its analysis of the 

issue of no significant hazards 

consideration, which is presented 

below: 

1. Does the proposed amendment involve 

a significant increase in the probability or 

consequences of an accident previously 

evaluated? 

Response: No. 

The proposed change to Vogtle Electric 

Generating Plant (VEGP) Technical 

Specification (TS) 3.3.1, ‘‘Reactor Trip 

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23818 Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 

System (RTS) Instrumentation,’’ Table 3.3.1– 

1, ‘‘Reactor Trip System Instrumentation’’ 

does not significantly increase the probability 

or consequences of an accident previously 

evaluated in the Updated Final Safety 

Analysis Report (UFSAR). The overall 

protection system performance will remain 

within the bounds of the accident analysis 

since there are no hardware changes. The 

design of the Reactor Trip System (RTS) 

instrumentation, specifically the positive 

range neutron flux high positive rate trip 

(PFRT) function, will be unaffected. The 

reactor protection system will continue to 

function in a manner consistent with the 

plant design basis. All design, material, and 

construction standards that were applicable 

prior to the request are maintained. 

The proposed change adds an additional 

surveillance requirement to assure that the 

PFRT is verified to be consistent with the 

safety analysis and licensing basis. In this 

specific case, a response time verification 

requirement will be added to the PFRT 

function. 

The proposed changes will not modify any 

system interface. The proposed changes will 

not affect the probability of any event 

initiators. There will be no degradation in the 

performance of or an increase in the number 

of challenges imposed on safety-related 

equipment assumed to function during an 

accident situation. There will be no change 

to normal plant operating parameters or 

accident mitigation performance. The 

proposed change will not alter any 

assumptions nor change any mitigation 

actions in the radiological consequences 

evaluations in the UFSAR. 

The proposed change does not adversely 

affect accident initiators or precursors nor 

alter the design assumptions, conditions, or 

configuration of the facility or the manner in 

which the plant is operated and maintained. 

The proposed changes do not alter nor 

prevent the ability of SSCs from performing 

their intended function to mitigate the 

consequences of an initiating event within 

the assumed acceptance limits. The proposed 

change is consistent with the safety analyses 

assumptions and resultant consequences. 

The RCS overpressure limit listed in 

Specification 2.1.2 of the VEGP Technical 

Specifications (i.e., 2735 psig) is not violated. 

2. Does the proposed amendment create 

the possibility of a new or different kind of 

accident from any previously evaluated? 

Response: No. 

There are no hardware changes nor are 

there any changes in the method by which 

any safety related plant system performs its 

safety function. This change will not affect 

the normal method of plant operation nor 

change any operating parameters. 

No performance requirements will be 

affected; however, the proposed change adds 

an additional surveillance requirement. The 

additional surveillance requirement is 

consistent with assumptions made in the 

safety analyses and licensing basis. 

No new accident scenarios, transient 

precursors, failure mechanisms, or limiting 

single failures are introduced as a result of 

this change. There will be no adverse effect 

or challenges imposed on any safety-related 

system as a result of this change. 

Therefore, the proposed change does not 

create the possibility of a new or different 

accident from any accident previously 

evaluated. 

3. Does the proposed amendment involve 

a significant reduction in a margin of safety? 

Response: No. 

The proposed change does not affect the 

acceptance criteria for any analyzed event 

nor is there a change to any Safety Limits. 

There will be no effect on the manner in 

which Safety Limits or Limiting Conditions 

of Operations are determined, nor will there 

be any effect on those plant systems 

necessary to assure the accomplishment of 

protection functions. 

This change is consistent with the 

assumptions made in the safety analyses. The 

addition of a surveillance requirement 

increases the margin of safety by assuring 

that the associated safety analysis 

assumption on the PFRT response time is 

verified. 

Therefore, the proposed change does not 

involve a significant reduction in the margin 

of safety. 

Based on the above, SNC concludes that 

the proposed amendment does not involve a 

significant hazards consideration under the 

standard set forth in 10 CFR 50.92(c), and, 

accordingly, a finding of ‘‘no significant 

hazards consideration’’ is justified. 

The NRC staff has reviewed the 

licensee’s analysis and, based on this 

review, it appears that the three 

standards of 10 CFR 50.92(c) are 

satisfied. Therefore, the NRC staff 

proposes to determine that the 

amendment request involves no 

significant hazards consideration. 

Previously Published Notices of 

Consideration of Issuance of 

Amendments to Facility Operating 

Licenses, Proposed No Significant 

Hazards Consideration Determination, 

and Opportunity for a Hearing 

The following notices were previously 

published as separate individual 

notices. The notice content was the 

same as above. They were published as 

individual notices either because time 

did not allow the Commission to wait 

for this biweekly notice or because the 

action involved exigent circumstances. 

They are repeated here because the 

biweekly notice lists all amendments 

issued or proposed to be issued 

involving no significant hazards 

consideration. 

For details, see the individual notice 

in the Federal Register on the day and 

page cited. This notice does not extend 

the notice period of the original notice. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–255, Palisades Nuclear 

Plant, Van Buren County, Michigan 

Date of amendment request: March 

31, 2010. 

Brief description of amendment 

request: The proposed amendment 

would add new license condition 2.C(4) 

stating that performance of Technical 

Specification surveillance requirement 

3.1.4.3, which verifies control rod 

freedom of movement, is not required 

for control rod drive 22 during cycle 21 

until the next entry into Mode 3 in a 

maintenance or refueling outage, 

whichever is earlier. 

Date of publication of individual 

notice in Federal Register: April 14, 

2010 (75 FR 19428). 

Expiration date of individual notice: 

June 13, 2010. 

Union Electric Company, Docket No. 

50–483, Callaway Plant, Unit 1, 

Callaway County, Missouri 

Date of amendment request: March 

29, 2010, as supplemented by letter 

dated March 29, 2010. 

Brief description of amendment 

request: The proposed amendment 

would revise the Technical 

Specification (TS) 3.3.2, ‘‘Engineered 

Safety Feature Actuation System 

(ESFAS) Instrumentation,’’ regarding 

function 6.g in TS Table 3.3.2–1. 

Function 6.g provides an auxiliary 

feedwater (AFW) start signal that is 

provided to the motor-driven AFW 

pumps in the event of a trip of both 

turbine-driven main feedwater pumps. 

The changes would revise Condition J 

for ESFAS instrumentation function 6.g 

to read, ‘‘One or more Main Feedwater 

Pumps trip channel(s) inoperable.’’ The 

licensee will make corresponding 

changes to Required Action J.1 and the 

Note above Required Actions J.1 and J.2 

for consistency with the revised 

Condition. 

Date of publication of individual 

notice in Federal Register: April 14, 

2010 (75 FR 19431). 

Expiration date of individual notice: 

April 28, 2010, for public comments; 

June 14, 2010, for hearing requests. 

Notice of Issuance of Amendments to 

Facility Operating Licenses 

During the period since publication of 

the last biweekly notice, the 

Commission has issued the following 

amendments. The Commission has 

determined for each of these 

amendments that the application 

complies with the standards and 

requirements of the Atomic Energy Act 

of 1954, as amended (the Act), and the 

Commission’s rules and regulations. 

The Commission has made appropriate 

findings as required by the Act and the 

Commission’s rules and regulations in 

10 CFR Chapter I, which are set forth in 

the license amendment. 

Notice of Consideration of Issuance of 

Amendment to Facility Operating 

License, Proposed No Significant 

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Federal Register / Vol. 75, No. 85 / Tuesday, May 4, 2010 / Notices 23819 

Hazards Consideration Determination, 

and Opportunity for A Hearing in 

connection with these actions was 

published in the Federal Register as 

indicated. 

Unless otherwise indicated, the 

Commission has determined that these 

amendments satisfy the criteria for 

categorical exclusion in accordance 

with 10 CFR 51.22. Therefore, pursuant 

to 10 CFR 51.22(b), no environmental 

impact statement or environmental 

assessment need be prepared for these 

amendments. If the Commission has 

prepared an environmental assessment 

under the special circumstances 

provision in 10 CFR 51.22(b) and has 

made a determination based on that 

assessment, it is so indicated. 

For further details with respect to the 

action see (1) the applications for 

amendment, (2) the amendment, and (3) 

the Commission’s related letter, Safety 

Evaluation and/or Environmental 

Assessment as indicated. All of these 

items are available for public inspection 

at the Commission’s Public Document 

Room (PDR), located at One White Flint 

North, Public File Area 01F21, 11555 

Rockville Pike (first floor), Rockville, 

Maryland. Publicly available records 

will be accessible from the Agencywide 

Documents Access and Management 

System (ADAMS) Public Electronic 

Reading Room on the internet at the 

NRC Web site, http://www.nrc.gov/ 

reading-rm/adams.html. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the PDR 

Reference staff at 1–(800) 397–4209, 

(301) 415–4737 or by e-mail to 

pdr.resource@nrc.gov. 

Entergy Nuclear Operations, Inc., 

Docket No. 50–333, James A. FitzPatrick 

Nuclear Power Plant, Oswego County, 

New York 

Date of application for amendment: 

November 23, 2009, as supplemented by 

letter dated February 5, 2010. 

Brief description of amendment: The 

amendment modified the Technical 

Specification (TS) 5.5.7, ‘‘Inservice 

Testing Program,’’ by replacing the 

references from the American Society of 

Mechanical Engineers (ASME) Boiler 

and Pressure Vessel Code to the current 

Code of Record, the ASME Operation 

and Maintenance Nuclear Power Plants 

Code (ASME OM Code), the Code of 

Record for the James A. FitzPatrick 

Nuclear Power Plant (JAFNPP) Inservice 

Testing (IST) Program. This is an 

administrative amendment to maintain 

the TS current with the NRC accepted 

Code of Record for JAFNPP IST 

Program. 

Date of issuance: April 12, 2010. 

Effective date: As of the date of 

issuance, and shall be implemented 

within 30 days. 

Amendment No.: 296. 

Renewed Facility Operating License 

No. DPR–59: The amendment revised 

the License and the Technical 

Specifications. 

Date of initial notice in Federal 

Register: January 26, 2010 (75 FR 4117). 

The February 5, 2010, supplement 

provided additional information that 

clarified the application, did not expand 

the scope of the application as originally 

noticed, and did not change the NRC 

staff’s original proposed no significant 

hazards consideration determination as 

published in the Federal Register. 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated April 12, 2010. 

No significant hazards consideration 

comments received: No. 

Exelon Generation Company, LLC, 

Docket Nos. STN 50–454 and STN 50– 

455, Byron Station (Byron), Unit Nos. 1 

and 2, Ogle County, Illinois 

Date of application for amendment: 

September 24, 2009, as supplemented 

by letters dated November 13, 2009; 

January 19, 2010; March 1, 2010; March 

9, 2010 (two letters); and March 19, 

2010. 

Brief description of amendment: The 

amendments adds a new Completion 

Time (CT) of 144 hours to restore a unitspecific essential service water train to 

operable status associated with the 

Limiting Condition for Operation for 

Technical Specification (TS) 3.7.8, 

‘‘Essential Service Water (SX) System.’’ 

The new CT will be used for 

maintenance during the Byron, Unit No. 

2, spring 2010, refueling outage. The 

licensee requested the new CT to 

replace two of the four SX pump suction 

isolation valves without having to 

shutdown Byron, Unit No. 1; 

maintenance history has shown that 

replacement of the SX pump suction 

isolation valves cannot be assured 

within the existing 72 hour CT window. 

Date of issuance: April 9, 2010. 

Effective date: As of the date of 

issuance and shall be implemented 

within 30 days. 

Amendment Nos.: Unit No. 1—168; 

Unit No. 2—168. 

Facility Operating License Nos. NPF– 

37 and NPF–66: The amendments revise 

the TSs and Licenses. 

Date of initial notice in Federal 

Register: December 1, 2009 (74 FR 

62835). 

The supplemental letters provided 

additional information that clarified the 

application, did not expand the scope of 

the application as originally noticed, 

and did not change the NRC staff’s 

original proposed no significant hazards 

consideration determination. 

The Commission’s related evaluation 

of the amendments is contained in a 

Safety Evaluation dated April 9, 2010. 

No significant hazards consideration 

comments received: No. 

R.E. Ginna Nuclear Power Plant, LLC, 

Docket No. 50–244, R.E. Ginna Nuclear 

Power Plant, Wayne County, New York 

Date of application for amendment: 

September 18, 2009. 

Brief description of amendment: The 

amendment revises Technical 

Specification (TS) 5.5.7, ‘‘Inservice 

Testing Program,’’ by incorporating TS 

Task Force Traveler (TSTF)-479, 

‘‘Changes to Reflect Revision of 10 CFR 

50.55a,’’ and TSTF–497, ‘‘Limit Inservice 

Testing Program SR [Surveillance 

Requirement] 3.0.2 Application to 

Frequencies of 2 Years or Less.’’ 

Specifically, the amendments (1) 

replace references to the American 

Society of Mechanical Engineers 

(ASME) Boiler and Pressure Vessel 

Code, Section XI with the ASME Code 

for Operation and Maintenance of 

Nuclear Power Plants for inservice 

testing activities, and (2) applies the 

extension allowance of SR 3.0.2 to other 

normal and accelerated inservice testing 

frequencies of 2 years or less that were 

not included in the frequencies listed in 

TS 5.5.7.a. 

Date of issuance: April 8, 2010. 

Effective date: As of the date of 

issuance to be implemented within 60 

days. 

Amendment No.: 110. 

Renewed Facility Operating License 

No. DPR–18: Amendment revised the 

License and Technical Specifications. 

Date of initial notice in Federal 

Register: November 3, 2009 (74 FR 

56887). 

The Commission’s related evaluation 

of the amendment is contained in a 

Safety Evaluation dated April 8, 2010. 

No significant hazards consideration 

comments received: No. 

Dated at Rockville, Maryland, this 22nd 

day of April 2010. 

For the Nuclear Regulatory Commission. 

Robert A. Nelson, 

Deputy Director, Division of Operating 

Reactor Licensing, Office of Nuclear Reactor 

Regulation. 

[FR Doc. 2010–10105 Filed 5–3–10; 8:45 am] 

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